In a nuclear reactor, the neutron flux and energy distribution are indications of various conditions, the knowledge of which is essential to the control of the reactor. Neutron dosimetry is carried out by fabricating dosimeters each containing a known number of atoms of a dosimeter target nuclide, placing the dosimeters in various locations within the reactor, and measuring the radioactivity imparted to the dosimeter by the neutron radiation. These dosimeters may contain fissile nuclides which, upon neutron irradiation, form radioactive fission fragments or stable nuclides which are transmuted into radionuclides upon neutron irradiation. The above nuclear processes are quantitative, so the radioactivity induced in the dosimeter target material during a given exposure time is proportional to the neutron flux and to the cross-section for the particular reaction. Since the cross-section is a function of neutron energy, the radiation count rate from exposed dosimeters is an indication of both the neutron flux and neutron energy. By disposing dosimeters of various target materials in various locations throughout the reactor, a profile or core characterization of neutron energy and flux can be calculated.
Suitable materials for dosimeter targets may be any nuclide which generates a radionuclide upon neutron irradiation. Any fissile material is suitable because these materials undergo fission upon exposure to neutrons, producing characteristic radioactive fission fragments. Most stable nuclides are also suitable as dosimeter targets as they generate radioactive activation products upon neutron irradiation.
In order to provide an accurate and reproducible representation of the neutron flux and energy within a reactor, each dosimeter should be of uniform thickness and contain a known number of atoms of the target nuclide. In addition, the target nuclide should be uniformly distributed throughout the dosimeter.